Techniques for on-demand production of medical isotopes such as mo-99/tc-99m and radioactive iodine isotopes including i-131

ABSTRACT

A system for radioisotope production uses fast-neutron-caused fission of depleted or naturally occurring uranium targets in an irradiation chamber. Fast fission can be enhanced by having neutrons encountering the target undergo scattering or reflection to increase each neutron&#39;s probability of causing fission (n, f) reactions in U-238. The U-238 can be deployed as one or more layers sandwiched between layers of neutron-reflecting material, or as rods surrounded by neutron-reflecting material. The gaseous fission products can be withdrawn from the irradiation chamber on a continuous basis, and the radioactive iodine isotopes (including I-131) extracted.

CROSS-REFERENCES TO RELATED APPLICATIONS

This application is a continuation of U.S. patent application Ser. No.14/709,408 filed May 11, 2015, published as U.S. Patent Publication No.2015/0243395, which is a divisional of U.S. patent application Ser. No.12/944,634 filed Nov. 11, 2010 and issued as U.S. Pat. No. 9,047,997,for “Techniques for On-Demand Production of Medical Isotopes Such asMo-99/Tc-99m and Radioactive Iodine Isotopes Including I-131.” Thisapplication also claims priority to the following U.S. patentapplications: U.S. Provisional Patent Application No. 61/260,585 filedNov. 12, 2009 for “Medical Isotope Production on Demand”; U.S.Provisional Patent Application No. 61/265,383 filed Dec. 1, 2009 for“System for On-Demand Production of I-131”; and U.S. Provisional PatentApplication No. 61/405,605 filed Oct. 21, 2010 for “Techniques forOn-Demand Production of Medical Isotopes Such as Mo-99/Tc-99m.” Thisapplication is also related to U.S. Pat. No. 8,989,335 issued Mar. 24,2015, entitled “Techniques for On-Demand Production of MedicalRadioactive Iodine Isotopes Including I-131.” The entire disclosures ofall the above mentioned applications, including all appendices andattachments, are hereby incorporated by reference for all purposes.

BACKGROUND OF THE INVENTION

The present invention relates generally to the generation of unstable,i.e., radioactive, nuclear isotopes (often referred to asradioisotopes), and more particularly to techniques for generatingmedical isotopes such as molybdenum-99 (Mo-99) and its decay daughtertechnetium-99m (Tc-99m), and radioactive iodine such as iodine-131(I-131).

Radioisotopes, in very small doses, are widely used in clinical therapy(radiation treatments) for such diseases as cancer and hyperthyroidism,as well as diagnostics using the ability to image regions whereradioisotopes concentrate in the subject's body. Currently, nearly 80%of all nuclear imaging procedures utilize Tc-99m, making it a veryimportant isotope for diagnostic medicine. Molecules and proteins thatconcentrate in specific areas of the body can be tagged with Tc-99m,which decays to a ground state through emission of a low energygamma-ray, and observed from outside the body using gamma-ray cameras ordetectors. This method allows “active” areas, or regions where theTc-99m tagged compound concentrates, to be observed in 3-D from outsidethe body.

With a high demand for medical procedures involving the use of Tc-99mand I-131, a demand that is only expected to increase as the U.S.population ages, reliability of the Tc-99m and I-131 supplies iscritical. A major obstacle to a reliable source is the fact that 100% ofthe U.S. supply is imported from foreign reactors. The U.S. supply issourced almost entirely from the NRU Reactor in Canada (Chalk River) andthe HFR in the Netherlands, and both reactors are over 40 years old. Therapid decay of the Mo-99 means that product must be shipped and usedimmediately with no long term storage possible. Any interruptions insupply, even brief periods such as a reactor shutting down formaintenance, can cause shortages and patient treatment delays. Realshortages have occurred as recently as 2007 and 2008 when the NRUReactor and HFR, respectively, were shut down for a period of time.

Traditional methods rely on thermal fission of targets made of highlyenriched uranium (HEU). HEU is uranium that has been processed togreatly increase the percentage of fissionable U-235 above theapproximately 0.7% level, found in naturally occurring uranium, tolevels above 93%. Thermal fission refers to the irradiation of a targetby low-energy (“thermal”) neutrons, causing fission to occur. Currently,the U.S. exports more than 50 kg of HEU having more than 93% U-235 to atleast five foreign nuclear reactors for irradiation and extraction ofthe Mo-99/Tc-99m and other medical isotopes. The proliferation potentialand hazards associated with shipping fresh HEU and spent HEU areobvious, but current production of Mo-99/Tc-99m relies on this process.

The spent HEU target material is also a threat because only between 1-3%of the U-235 in the HEU target is burned up and the remaining targetmaterial can still contain 92% enriched U-235. Also, because of the lowburn-up, after three-year storage, the HEU target materials can beessentially contact handled, meaning that due to its relatively low burnup, the amount of long-lived fission products in the spent HEU targetmaterial are minimal. The spent HEU target can be handled and processedrelatively easy with minimal shielding materials to protect theproliferators.

Alternative techniques have been proposed, but they are thought to besignificantly less cost effective and many technical challenges remain.One such proposal is to transition to a lower level of enrichment of theU-235 target (LEU), say below 20% U-235, but this still presents thesame problems as HEU, including the need for a nuclear reactor. Anothermethod proposed is to utilize neutron capture in Mo-98, which can bemined as ore. However, natural molybdenum contains on the order of 24.1%Mo-98, so targets are likely to require enrichment prior to irradiation.

Other techniques proposed for production of Mo-99 include causing (p,2n) or (p, pn) reactions in an Mo-100 target using a proton accelerator(cyclotron). Again, natural molybdenum contains on the order of 9.64% ofMo-100, so targets for cyclotrons are also likely to require enrichmentprior to irradiation. It has also been proposed to cause (γ, f)reactions on U-235 or LEU, or U-238 or a combination of all threematerials via bremsstrahlung radiation produced from a high-energyelectron accelerator.

The proposed alternative methods using particle accelerators all havesimilar problems:

-   -   They all require large and enriched isotopic targets.    -   They all require heat removal from the targets during        irradiation, which represents a technical challenge.    -   The Mo-99 produced must be purified to remove unused molybdenum        isotopes and other fission products and activation by-products.    -   They all require development of fast dissolution methods for the        metallic targets.    -   Treatment and disposal of the waste fission products and waste        uranium present significant challenges (for LEU).

SUMMARY OF THE INVENTION

Embodiments of the present invention provide techniques for theproduction of radioisotopes. Radioisotopes that are useful in the fieldof medicine are sometimes referred to as medical isotopes, although somestable isotopes have potential medical uses and are sometimes referredto as stable medical isotopes. The techniques provided by the inventionovercome at least some of the problems discussed above. Embodiments donot rely on a nuclear reactor far from the delivery site but can beimplemented as relatively small stand-alone devices that can be widelydistributed.

In describing embodiments of the invention, the following terms aresometimes used:

-   -   Non-enriched uranium (“NEU”);    -   Neutron-reflecting material;    -   Fast neutrons;    -   Fast neutron fission reactions; and    -   Neutron generator.

As will now be discussed, these terms are defined broadly.

The term “non-enriched uranium” (“NEU”) is intended to cover naturallyoccurring uranium or depleted uranium, in addition to any uranium thatcontains at least as much U-238 as naturally occurring uranium (99.27%)and no more U-235 than naturally occurring uranium (0.72%). Depleteduranium is normally understood to mean uranium that has less than thenaturally occurring amount of U-235 (0.72%), but depleted uranium thatis used for commercial and military purposes more commonly has less than0.3% U-235.

The definition of NEU is not limited to any form of the uranium, so longas the isotope content meets the above criteria. The NEU material, alsoreferred to as the NEU feedstock or the NEU target, can be in the formof bulk solid material, crushed solid material, metallic shavings,metallic filings, sintered pellets, liquid solutions, molten salts,molten alloys, or slurries. The NEU, whatever its form, can also bemixed with other material that is compatible with the intended use.

The term “neutron-reflecting material” is intended to cover materialthat reflects or scatters neutrons. While it is preferred that thescattering be elastic, or largely so, this is not necessary for thedefinition. Further, while some of the embodiments useneutron-reflecting material formed into solid structural shapes such asplates, spherical shells, cylindrical shells, tubes, and the like, theterm is intended to cover material that includes small particles such aspowders, pellets, shavings, filings, and the like.

The term “fast neutron” is often used to distinguish thermal neutrons,which Wikipedia characterizes as having energies of “of about 0.025 eV.”Wikipedia also characterizes fast neutrons as having energies “greaterthan 1 eV, 0.1 MeV or approximately 1 MeV, depending on the definition.”For present purposes, the term “fast neutron” will mean a neutron withan energy above 800 keV (i.e., 0.8 MeV), which is a threshold forfission in U-238. However, embodiments of the present invention can useneutrons of higher energies, say 10-20 MeV, or possibly 12-16 MeV.Higher-energy neutrons, say in the 20-100 MeV range, can also be used.

The term “fast neutron fission” is intended to cover the fissionreactions that are caused by neutrons with energies that are above thethreshold of 800 keV. The reaction representation (n, f) is used forsimplicity.

The term “neutron generator” is intended to cover a wide range ofdevices and processes for generating neutrons of the desired energies.Wikipedia defines neutron generators as follows:

-   -   Neutron generators are neutron source devices which contain        compact linear accelerators and that produce neutrons by fusing        isotopes of hydrogen together. The fusion reactions take place        in these devices by accelerating either deuterium, tritium, or a        mixture of these two isotopes into a metal hydride target which        also contains either deuterium, tritium or a mixture. Fusion of        deuterium atoms (D+D) results in the formation of a He-3 ion and        a neutron with a kinetic energy of approximately 2.45 MeV.        Fusion of a deuterium and a tritium atom (D+T) results in the        formation of a He-4 ion and a neutron with a kinetic energy of        approximately 14.1 MeV. Fusion of a triton and a tritium atom        (T+T) results in the formation of a He-4 ion and two neutrons.        These two neutrons can have an energy range from below 0.1 eV to        ˜9.33 MeV.

As used in this application, however, the term “neutron generator” isdefined more broadly to include any device that would provide asufficient number of neutrons of the desired energies. This couldinclude, for example, but is not limited to, the following.

-   -   A dense plasma focusing device can use deuterium or tritium        plasma to produce 2.45 MeV neutrons, 14.1 MeV neutrons, or        neutrons covering a broad spectrum (below 0.1 eV to ˜9.33 MeV).    -   An electron accelerator can be used to send high energy        electrons from an electron beam onto a converter material, e.g.,        tantalum (Ta), tungsten (W), etc., thereby converting the        electron energy into bremsstrahlung radiation. This        bremsstrahlung radiation can then be used to interact with        neutron-rich materials to produce neutrons via (γ, n)        interactions. For example, irradiating the beryllium isotope        Be-9 with γ rays can produce a beryllium isotope with a lower        atomic mass and one or two neutrons (the reactions being denoted        Be-9(γ, n)Be-8, or Be-9(γ, 2n)Be-7).    -   A proton accelerator such as a cyclotron can be used to send        high energy protons into materials such as carbon, beryllium, or        lithium, for example, to produce neutrons via C-12(p, n)N-12, or        Be-9(p, n)B-9 reactions, for example.

In short, embodiments of the present invention use fast-neutron-causedfission of depleted or naturally occurring uranium targets in anirradiation chamber. A generic term for such uranium is non-enricheduranium (“NEU”). U-238, is fissionable in that it can be made to fissionwhen struck by fast neutrons, i.e., neutrons having energies above afission threshold. It is not fissile in that it cannot sustain a chainreaction. This is because when U-238 undergoes fission, neutronsresulting from the fission are generally below the energy threshold tocause more U-238 fission.

The purpose of causing fission is to generate and extract fissionproducts that are, or decay to, desired radioisotopes. Embodiments willbe described in the context of extracting Mo-99/Tc-99m, but this isexemplary. Since the fission products include radioactive iodineisotopes, some embodiments can also extract radioactive iodine isotopesas well.

In this application, the term “fission product” will be used as setforth in the NRC glossary(http://www.nrc.gov/reading-rm/basic-ref/glossary.html), which defines“fission products” as “Mlle nuclei (fission fragments) formed by thefission of heavy elements, plus the nuclide formed by the fissionfragments' radioactive decay.” Thus the term is used more broadly than adefinition that would cover only the nuclei resulting directly from thefission reaction.

This interpretation is consistent with the glossary athttp://www.nuclearglossary.com (“The Language of the Nucleus”), whichdefines “fission product” as “[a]residual nucleus formed in fission,including fission fragments and their decay daughters.” The term“fission fragment” is defined as “[a]nucleus formed as a direct resultof fission. Fission products formed by the decay of these nuclides arenot included.” The term “primary fission product” is said to be asynonym for “fission fragment.”

Embodiments of the present invention operate to enhance fast fission inNEU targets by having neutrons undergo scattering or reflection afterpassing through a region of NEU. This is accomplished by having thetarget material interspersed with what is referred to as“neutron-reflecting” material, which reflects or scatters neutrons sothat the neutrons travel a longer path before leaving the targetmaterial. This provides more opportunities for the neutrons to causefission reactions with the NEU target material.

Thus, a given neutron tends to have multiple interactions (e.g.,scattering events) with U-238 nuclei before it causes a fission reactionor leaves the region or regions occupied by NEU target material withoutbeing absorbed by the U-238. After a number of scattering events withinthe NEU target material, a neutron's energy will drop below the fissionthreshold. The target material can be interspersed withneutron-reflecting material according to a number of differentgeometrical arrangements.

The particular form of feedstock typically depends on the geometry ofthe irradiation chamber, the characteristics of the fission productsthat are to be extracted, and the manner of extracting the fissionproducts. A preferred feedstock is in the form of depleted uranium.Depleted uranium is a byproduct of uranium enrichment and contains over99.7% of U-238 as compared to natural uranium, which contains about99.3% of U-238.

As well as maintaining the neutron energy above a fast fissionthreshold, it is preferred to maintain a sufficiently high neutronenergy to minimize neutron absorption by U-238. When a neutron withslower energy is captured by U-238, Pu-239 is produced after subsequentdecay of the excited U-238 atom. Using fast neutron sources instead, theprobability of fission of U-238 becomes orders of magnitude higher thanthe probability of capture, resulting in greatly reduced production ofPu-239.

In one set of embodiments, the NEU and neutron-reflecting material areformed as alternating layers of NEU and neutron-reflecting material. Thelayers can take the form of spherical shells, cylindrical shells, flatplates and the like, with a fast neutron generator disposed near thecenter. In these embodiments, the irradiation chamber is generallyspherical, generally cylindrical, or generally rectangular. Othergeometries such as polygonal cylinders and polyhedrons are alsopossible, and may allow easier fabrication.

In another set of embodiments, the NEU occupies a plurality of parallelelongate regions with each region surrounded by neutron-reflectingmaterial. The surrounding neutron-reflecting material can be formed as aplurality of tubes, and solid rods or crushed NEU can be disposed in thetubes. The NEU and surrounding neutron-reflecting material can occupy acylindrical region with a hollow center for the fast neutron generator.Thus, in this set of embodiments, the irradiation chamber is generallycylindrical (circular or polygonal base).

In other embodiments, the NEU and neutron-reflecting material can bothbe formed as relatively small objects (say a few centimeters in size),and mixed in solid form or in a slurry. The slurry can be circulated intubes surrounding the neutron generator. Embodiments where the NEU is inmolten form, whether or not interspersed with neutron-reflectingmaterial, can also be circulated. Circulating the NEU results in evenirradiation of all the NEU in the chamber.

In an aspect of the invention, a method for producing radioisotopescomprises introducing NEU material into a an irradiation chamber,irradiating the NEU material with neutrons having energies above 800 keVto cause fast fission reactions to occur in the NEU material andgenerate fission products, and extracting the fission products from theNEU material. The irradiation chamber has one or more walls formed ofneutron-reflecting material, and at least some neutrons from theirradiating are reflected from at least one of the one or more walls,thereby increasing the path length over which those neutrons are in theNEU material. The increased path length increases the probability thatthose neutrons in the NEU material will cause fast fission reactions.

In some embodiments, the NEU material in the irradiation chamberoccupies a single spatially contiguous region, while in otherembodiments, the NEU material in the irradiation chamber occupiesmultiple spatially disjoint regions. The one or more walls formed ofneutron-reflecting material can comprise at least one internal wall ofthe irradiation chamber, or an outer wall that surrounds all the NEUmaterial in the irradiation chamber, or both.

In another aspect of the invention, a method for producing radioisotopescomprises providing a volume of NEU material, interspersing the NEUmaterial with neutron-reflecting material, surrounding the volume of NEUmaterial with additional neutron-reflecting material, surrounding theadditional neutron-reflecting material with neutron-absorbing material,and irradiating the NEU material with neutrons having energies above afission threshold to cause fast fission reactions to occur in the NEUmaterial and generate fission products.

For at least some neutrons, the neutron-reflecting material prolongs thetime that those neutrons remain within the volume of NEU material,thereby increasing the number of fast fission reactions caused by thoseneutrons before those neutrons encounter the neutron-absorbing material.The method can also include extracting the fission products from the NEUmaterial. Depending on the reactions and the products, the extractionmay or may not require removing the NEU material from the irradiationchamber.

In another aspect of the invention, apparatus for producingradioisotopes comprises a fast neutron generator and a plurality ofspaced shells made of neutron-reflecting material. The shells surroundthe neutron generator and include an outermost shell, and the spacingbetween adjacent shells provides a number of regions configured toreceive NEU for irradiation by neutrons generated by the neutrongenerator. The outermost shell can be thicker than the remaining shells.

In another aspect of the invention, apparatus for producingradioisotopes comprises a fast neutron generator, an irradiation chamberhaving one or more regions into which NEU can be introduced, and one ormore neutron-reflecting regions disposed in or around the irradiationchamber. The one or more neutron-reflecting regions are configured toincrease the path length traveled by at least some neutrons from theneutron generator before those neutrons leave the irradiation chamber.

In some embodiments, the apparatus can also comprise an outercontainment vessel having one or more walls made of neutron-absorbingmaterial to absorb neutrons passing out of the outermost shell. Thewalls of the outer containment vessel can be spaced from the outermostshell to limit the likelihood that neutrons scattered or reflected fromthe walls of the outer containment vessel will encounter the outermostshell.

In some embodiments, the irradiation chamber is of sphericalconfiguration, the one or more neutron-reflecting regions include aplurality of spaced concentric spherical shells, including an outermostshell, of neutron-reflecting material, the shells surround the neutrongenerator, and the space between adjacent shells defines at least one ofthe one or more regions into which NEU can be introduced.

In some embodiments, the irradiation chamber is of cylindricalconfiguration with an outer wall having a portion formed as acylindrical shell, the one or more neutron-reflecting regions includeone or more tubes of neutron-reflecting material, and the bores of theone or more tubes define at least one of the one or more regions intowhich NEU can be introduced.

In some embodiments, the irradiation chamber is of cylindricalconfiguration with an outer wall having a portion formed as acylindrical shell, the one or more neutron-reflecting regions include aplurality of spaced coaxial cylindrical shells of neutron-reflectingmaterial, and the space between the cylindrical shells ofneutron-reflecting material defines at least one of the one or moreregions into which NEU can be introduced.

In some embodiments, the irradiation chamber is of rectangularconfiguration with an outer wall having a portion formed as arectangular shell, the one or more neutron-reflecting regions include aplurality of spaced plates or rectangular shells of neutron-reflectingmaterial, and the space between the plates or rectangular shells ofneutron-reflecting material defines at least one of the one or moreregions into which NEU can be introduced.

A further understanding of the nature and advantages of the presentinvention may be realized by reference to the remaining portions of thespecification and the drawings, which are intended to be exemplary andnot limiting.

BRIEF DESCRIPTION OF THE DRAWINGS

FIG. 1 is a stylized high-level schematic of a radioisotope generatoraccording to an embodiment of the present invention.

FIG. 2 shows the decay products of Mo-99 after it has been generated byfast fission of U-238.

FIG. 3 is a graph of U-238's fission cross section and neutron capturecross section as functions of neutron energy.

FIG. 4 is a simplified oblique view of a radioisotope generator having aspherical irradiation chamber according to an embodiment of the presentinvention.

FIG. 5 is a simplified oblique view of a radioisotope generator having acylindrical irradiation chamber according to an embodiment of thepresent invention.

FIG. 6 is a simplified cross-sectional view that applies to either ofthe radioisotope generators shown in FIGS. 4 and 5 and is taken alongline 6-6 in FIGS. 4 and 5.

FIG. 7 is a simplified oblique view of a radioisotope generator having arectangular-prism-shaped irradiation chamber according to an embodimentof the present invention.

FIG. 8 is a simplified cross-sectional view of the radioisotopegenerator shown in FIG. 7 and is taken along line 8-8 in FIG. 7.

FIG. 9 is a simplified cross-sectional view of a radioisotope generatorwhere the NEU is in the form of rods.

FIG. 10 is a process schematic for the generation of radioisotopesaccording to an embodiment of the present invention.

FIG. 11A is a perspective view of a radioisotope generator according toa specific embodiment having a cylindrical irradiation chamber withparallel NEU rods.

FIG. 11B is a perspective, cutaway view of the radioisotope generatorshown in FIG. 11A with the cover and the outer cylinder removed.

FIG. 11C is a perspective cutaway view of the radioisotope generatorshown in FIG. 11A taken from a different view to show additionaldetails.

FIG. 12 shows the decay products of I-131 after it has been generated byfast fission of U-238.

FIG. 13 is a simplified cross-sectional view of a radioisotope generatorsuch as that shown in the sectional view of FIG. 6, with additionaldetails relating to the circulation of gas through the irradiationchamber.

DESCRIPTION OF SPECIFIC EMBODIMENTS Overview of Embodiments

FIG. 1 is a stylized high-level schematic of a radioisotope generator 10according to an embodiment of the present invention, and is used toillustrate salient features that can be viewed as generic to the variousembodiments discussed below. The main components of radioisotopegenerator 10 include an irradiation chamber 15, a fast neutron generator20, and, for preferred embodiments, one or more neutron-reflectingelements 25. The irradiation chamber is configured to accept a charge ofwhat is referred to as non-enriched uranium (“NEU”).

As mentioned above, the NEU can be in any suitable form, includingelemental metal, salt, alloy, molten salt, molten alloy, slurry, orother mixture, and can assume any one of a number of shapes and states,as will be described below. For purposes of generality, the NEU is shownas a plurality of arbitrary-shaped bodies 30 (stippled for clarity). Theirradiation chamber is generally provided with mechanisms forintroducing NEU into, and removing NEU from, the irradiation chamber,such as one or more fill ports 35 a and one or more empty (drain) ports35 b. In some embodiments, it is desired to flow gas through theirradiation chamber, and to this end the chamber can be provided withone or more gas inlet ports 40 a and one or more gas outlet ports 40 b.

FIG. 1 also shows a neutron-absorbing outer containment chamber 45surrounding the irradiation chamber. In representative embodiments, thisouter containment chamber is an entirely separate structure from theirradiation chamber and can be built on site prior to installation ofthe radioisotope generator. Containment chamber 45 can be built as aconcrete vault or bunker and lined with special neutron-absorbingmaterial such as borated polyethylene. It is preferred that the outercontainment chamber be significantly larger than the irradiation chamberso that the irradiation chamber subtends a relatively small solid angleat any point on walls of the outer containment chamber. This tends toreduce the likelihood of any slow neutrons scattered by the outercontainment chamber walls getting back inside the irradiation chamber.

The walls of irradiation chamber 15 preferably include at least a layerof neutron-reflecting material, and so the combination of irradiationchamber 15's layer of neutron-reflecting elements, andneutron-reflecting elements 25 can be considered to define one or moreneutron-reflecting regions disposed in or around the irradiationchamber. Some of the embodiments to be described below include separateneutron-reflecting elements inside the chamber, which may have theeffect of subdividing the chamber into disjoint regions occupied by NEUmaterial, while other embodiments provide a single contiguous regionoccupied by NEU material.

For generality, FIG. 1 shows the NEU as separate disjoint bodies, butthe case of a single contiguous body of NEU material should beconsidered to fall within the scope of the invention. In keeping withthe schematic nature of FIG. 1, the neutron-reflecting elements and theNEU bodies are shown as sparsely populating the irradiation chamber. Inmost embodiments, however, the regions of NEU material will generallyfill hollow regions within the irradiation chamber so as to besurrounded by neutron-reflecting material. However, this does notpreclude their being hollow portions of the chamber having neitherneutron-reflecting material nor NEU material.

Embodiments of the present invention are not limited to any particulartype of neutron generator. Representative systems can use a portable andcompact accelerator that can accelerate and direct charged particles, orneutral particles, or deuterons, or tritons to targets that can be usedto produce neutrons with energies that are above the 800 keV fastneutron fission threshold energy of U-238. The target materials can beelements, compounds, or solutions. Other target materials also cancontain materials or compounds that are enriched with tritium atoms.Specifically, the fusing of deuterium atoms with tritium atoms produces14.1-MeV neutrons which are used to cause fast fission reactions in theuranium atoms. When tritons are accelerated and fused with targetsenriched with tritium atoms, the produced neutrons can encompassenergies from below 0.1 eV to ˜9.3 MeV.

Suitable neutron generators are commercially available, for example fromAdelphi Technology Inc., 2003 East Bayshore Rd, Redwood City, Calif.94063, Halliburton, 10200 Bellaire Blvd., Houston, Tex. 77072, andSchlumberger Technology, 300 Schlumberger Dr., Sugar Land, Tex. 77478.

The operation of radioisotope generator 10 can be summarized as follows.Neutron generator 20 provides neutrons above a fission threshold forU-238. The purpose of causing fission is to generate and extract one ormore fission products that are, or decay to, desired radioisotopes.

FIG. 2 shows the decay products of Mo-99 (Z=42), which is one of theprimary fission products, and is also produced through the decay chain.Mo-99 constitutes ˜6% of the fission yield. Mo-99 will decay, with ahalf-life of 66 hours, to Tc-99 (12.5%) and Tc-99's metastable isomerTc-99m (87.5%) (both Z=43). Both Mo-99 and Tc-99m can be recovered withchemical extraction processes and/or electrochemical separators.Embodiments will be described in the context of extracting Mo-99/Tc-99m.The neutrons are preferably maintained above energies where theneutron-absorption cross section for U-238 remains negligible. Thus, theNEU target does not breed plutonium (Pu-239). Other fission productsinclude I-129, I-131, I-132, and I-133, which can also be recovered withchemical extraction processes and/or electrochemical separators.

The purpose, and operation, of the neutron-reflecting material is toincrease the path length traveled by at least some neutrons from theneutron generator before those neutrons leave the irradiation chamber.Thus, as a neutron exits a region of NEU material, possibly havingundergone one or more scattering interactions therein, it is reflectedor scattered to enhance the likelihood that it will encounter additionalNEU material and therefore have an increased chance to initiate afission reaction. Iron is an example of an element that can act asneutron-reflecting material. In specific embodiments, stainless steel isused both for its structural and neutron-reflecting properties. Forneutron energies above 800 keV, each scattering event with iron causesthe neutron to lose about 0.56 MeV in energy. Thus, after a sufficientnumber of scattering events (depending on initial energy), the neutronwill fall below the fast fission threshold.

Overview of Relevant Properties of Uranium and Fission Reactions

The literature concerning the properties of uranium and the physics ofnuclear fission is vast, and is well understood by those skilled in thenuclear physics and engineering fields. For the sake of completeness, ashort overview of the relevant aspects of this vast store of knowledgewill be outlined to provide context for the description of embodimentsof the present invention.

As discussed above, embodiments of the present invention use depleted ornaturally occurring (i.e., non-enriched uranium or NEU) targets.Naturally occurring uranium is about 99.27% U-238, 0.72% U-235, and0.0055% U-234. Depleted uranium is the by-product of the process ofenriching naturally occurring uranium to achieve a higher proportion ofU-235. Thus the depleted uranium contains significantly less U-235 andU-234 than natural uranium (say less than a third as much). There is nofundamental reason why embodiments of the present invention could notuse pure U-238, but as a practical matter, that would be much moreexpensive.

Further, as discussed above, embodiments of the present inventionirradiate NEU, largely containing U-238, with fast neutrons to causefission. The most common nuclear reactors, on the other hand, irradiateU-235 with thermal neutrons to cause fission. Both U-235 and U-238 willundergo fission when struck with fast neutrons, but the characteristicsare different.

First is the difference in fission cross section as a function ofneutron energy for the two isotopes. U-235's fission cross section forfast neutrons is at least 250 times lower than for thermal neutrons.That is the reason why a nuclear reactor is used to produce Mo-99 usingtargets enriched with U-235, since the neutrons within a nuclear reactorare typically thermalized. Thermal neutrons cannot cause fission inU-238.

Second is the nature of the reaction. U-235 is fissile, meaning that itcan sustain a chain reaction when a critical mass is present, since theneutrons resulting from the fission reactions have energies whereU-235's fission cross section is high. U-238, on the other hand, whilefissionable, is not fissile. While U-238 can be made to fission whenstruck with fast neutrons, most of the neutrons resulting from thefission reactions have energies that are not sufficiently high to causeadditional U-238 fission.

FIG. 3 is a graph of U-238's fission cross section and neutron capturecross section as functions of neutron energy. This graph was generatedusing the tools provided at http://atom.kaeri.re.kr/cgi-bin/endfplot.pl.U-238 also has a reaction path where it absorbs a neutron and, after twobeta decays, becomes Pu-239. While this reaction path may be desirablein breeder reactors, embodiments of the invention seek to reduce theprobability of neutron absorption by maintaining neutron energies in arange where the cross section for fission far exceeds the cross sectionfor neutron absorption. For example, as can be seen in the graph of FIG.3, the two cross sections are equal at a neutron energy of ˜1.3 MeV,while at 2.2 MeV, the fission cross section is ˜20 times larger.Accordingly, it is preferred to minimize or eliminate materials thatwould act as moderators.

As can be seen from the graph of U-238's fission cross section as afunction of neutron energy, the fission cross section plateaus between 1and 2 MeV, the neutron capture cross section is rapidly falling off inthis range of energies.

It should be understood that a fast neutron can transverse its entirepath through the NEU without causing a fission reaction. This fastneutron also can be scattered or reflected from U-238 nuclei. Theneutron-reflecting or scattering material is used to enhance theprobability that a fast neutron will ultimately interact with a U-238nucleus and cause a fission reaction before its energy drops below the800 keV threshold. Once this fast neutron interacts with U-238 andcauses a fission reaction, this neutron essentially is gone. There willbe 2-3 neutrons born after the fission reaction (prompt neutrons), andsome of these neutrons can cause some addition fission in the NEU iftheir energies can stay above the 800 keV threshold.

Overview of Representative Geometrical Configurations

FIG. 4 is a simplified oblique view of a radioisotope generator 10_(sphere) having a spherical irradiation chamber according to anembodiment of the present invention. Fill port 35 a is shown, but drainport 35 b would be hidden from view. The chamber is preferablyfabricated in at least two sections, which is signified by an equatorialline (the hidden half being shown in phantom). The equatorial line alsodenotes the intersection of a horizontal plane that is designated by asection line 6-6. For simplicity, gas inlet port(s) 40 a and gas outletport(s) 40 b are not shown.

FIG. 5 is a simplified oblique view of a radioisotope generator 10_(cyl) having a cylindrical irradiation chamber according to anembodiment of the present invention. Again, fill port 35 a is shown, butdrain port 35 b would be hidden from view. Also, gas inlet port(s) 40 aand gas outlet port(s) 40 b are not shown. Shown in phantom is theintersection of a horizontal plane that is designated by a section line(also 6-6). The cylindrical wall of irradiation chamber 15 need not befabricated in multiple sections, although it can be. In such a case, thephantom line designating the intersection of the horizontal plane couldbe drawn in solid on the visible portion of the cylindrical wall.

FIG. 6 is a simplified cross-sectional view that applies to either ofradioisotope generators 10 _(sphere) and 10 _(cyl), and is taken alongline 6-6 in FIGS. 4 and 5. For the spherical embodiment 10 _(sphere),neutron-reflecting elements 25 are formed as a plurality of spacedconcentric spherical shells that surround neutron generator 20. For thecylindrical embodiment 10 _(cyl), neutron-reflecting elements 25 areformed as a plurality of spaced coaxial cylindrical shells that surroundneutron generator 20.

In both embodiments, NEU bodies or regions 30 partially or fully occupythe spaces between neutron-reflecting shells 25. The drawing issimplified in that it doesn't show holes in the neutron-reflectingshells that allow NEU material introduced in one region to find its wayto other regions. Further, it is contemplated that there may bebulkheads, again not shown, that maintain the spacing between the shellsand provide additional structural strength.

For both embodiments, the outermost neutron-reflecting shell, which isshown as being thicker than the radially inward shells, at leastpartially defines irradiation chamber 15. While a plurality of regionsfor NEU are shown, some embodiments can have only one region (sphericalor cylindrical shell as the case may be). Furthermore, the outermostneutron-reflecting shell need not be thicker than the inner one or ones.

Radially outward of the outermost neutron-reflecting shell is abiological shield 50, which is used to block ionizing radiation such asalpha particles, electrons, and gamma rays that might leak out of theneutron-reflecting shell. Biological shield 50 also can be considered topartially define irradiation chamber 15. Such a shield can be made ofmaterials such as lead, iron, borated polyethylene, or a combination ofany or all of such materials.

While specific dimensions do not form a part of the invention, somerepresentative dimensions, or at least factors that can be considered inspecifying particular dimensions will be discussed. For example, the NEUcan be formed in a single layer on the order of 30 to 50 cm thick, or acombination of multiple layers on the order of 10 cm thick and separatedby stainless steel layers on the order of 0.5 cm thick. In this way,when the neutron energy falls below 1 MeV after multiple scatteringevents, the neutron will leak out of the outermost layer of the NEU. Adistance of containment chamber 45's walls from irradiation chamber 15of at least 1 meter and generally on the order of 2.5 meters.

FIG. 7 is a simplified oblique view of a radioisotope generator 10_(rect) having a rectangular-prism-shaped irradiation chamber accordingto an embodiment of the present invention. Again, fill port 35 a isshown, but drain port 35 b would be hidden from view. Also, gas inletport(s) 40 a and gas outlet port(s) 40 b are not shown. Shown in phantomis the intersection of a horizontal plane that is designated by asection line 8-8.

FIG. 8 is a simplified cross-sectional view of the radioisotopegenerator shown in FIG. 7 and is taken along line 8-8 in FIG. 7. In thisembodiment, neutron-reflecting elements 25 are formed as a plurality ofspaced parallel plates, some of which are cut away or shortened toprovide a cavity for neutron generator 20. NEU bodies or regions 30partially or fully occupy the spaces between neutron-reflecting plates25.

FIG. 9 is a simplified cross-sectional view of a radioisotope generator10 _(rod) where the NEU is in the form of rods 55, one of which is shownenlarged. As can be seen, each rod comprises a tubular shell ofneutron-reflecting material 25 surrounding a bore 44 containing NEU 30.The rods are shown as circular in cross section, but polygonal tubes canalso be used. A circular cross section is generally preferred since thatis generally the most common and would be more economical to manufactureand operate. A circular cross section has the additional advantages ofthe greatest structural integrity and efficiency for the propagation offast neutrons.

The tubes are shown disposed in a cylindrical irradiation chamber, butthere is no requirement. As schematically drawn, rods are laid out in anoctagonal array, and so an octagonal irradiation chamber could also beused. The rods could also be distributed along a set of concentriccircles so that their axes would lie in concentric cylindrical surfaces.

Process Overview

FIG. 10 is a process schematic for the generation of radioisotopesaccording to an exemplary embodiment of the present invention. Eachblock in the schematic represents an operation, but many of the blocksalso represent a physical piece of apparatus. The connector linesrepresent logical flow as well as material flow, so the figure can alsobe viewed as a flowchart.

The production can be considered to begin with providing an initialsupply of NEU (operation 60), which in this exemplary embodiment isdepleted uranium. This exemplary embodiment uses NEU in granular form,and so NEU is then subjected to a grinding operation 65 and a sortingoperation 70 that rejects undersized pieces of NEU. The pieces meetingthe desired size criteria are sent to an inventory 75 of NEU in asuitable form or state, and the undersized pieces are diverted to bypassthe inventory and are subjected to further processing as will bedescribed below. Grinding and sorting operations 65 and 70 can also beused for recycling NEU as will be described below.

For the sake of this exemplary embodiment, a suitable form would be NEUthat had been ground or crushed to “pebbles” of desired nominal size,say on the order of no less than 0.64 cm (¼ inch) in the smallestdimension and no more than 12.7 cm (½ inch) in the largest dimension.The NEU is loaded into the radioisotope generator, say by gravitythrough fill port(s) 35 a (not shown in FIG. 10), and irradiated for asuitable period of time with neutrons of a suitable energy (operation80). In one exemplary embodiment, the NEU is irradiated with 3.5×10¹³neutrons/second for 20 hours using 14.1 MeV neutrons. This provides abalance between production and decay. For the example of a single 30 cmspherical shell of NEU having an outer diameter of 182.88 cm (6 feet),the mass of the NEU would be on the order of 22,000 kg.

After irradiation, the NEU, which now contains fission products,including the desired fission products and other fission products, isremoved from the radioisotope generator, say by gravity through drainport(s) 35 b (not shown in FIG. 10), subjected to a radioisotoperecovery operation 85 to provide the desired fission product, in thiscase Mo-99. The recovered Mo-99 is then subjected to a quality controltesting operation 90, is used to charge a Tc-99m generator (operation95), and the Tc-99m generator is set for shipment to an end user(operation 100). By way of example, a single generator may contain 6curies and would be used to prepare a large number of individual patientdoses that might be on the order of 10-30 microcuries per patient.

The NEU from which the Mo-99 has been recovered is subjected to aseparate recovery operation 105 to remove the other fission products,some of which may be desirable radioisotopes, and is then subjected toyet another recovery operation 110 to recover the NEU for recycling. Aswill be described below, the recovery operations can use ionic liquids,and more specifically room-temperature ionic liquids (RTILs).

The recovered NEU provided by recovery operation 110 is returned to besubjected to grinding and sorting operations, which can be the samegrinding and sorting operations 65 and 70 used for the NEU that isoriginally provided to the system. As for the originally provided NEU,the sorting operation rejects undersized pieces of NEU. The piecesmeeting the desired size criteria are returned to the NEU inventory, andthe undersized pieces are diverted to bypass the inventory andirradiation chamber.

The irradiation and fission can also give rise to various fissionproducts in a gaseous state. These gaseous fission products includefission products that are themselves gases (e.g., xenon and krypton),and iodine (e.g., I-129, I-131, I-132, I-133, etc.), which is a solid,but easily sublimates. In some traditional systems using HEU, the HEUtarget elements are encapsulated. Thus, these gaseous fission productswould be trapped in the encapsulated target elements, and the gaseousfission products would be captured after irradiation in connection withthe Mo-99 recovery. Additionally, to the extent that gaseous fissionproducts leaked out of the target, the iodine would dissolve in thewater that acted as a coolant and moderator.

In this exemplary embodiment, the gaseous fission products (i.e.,fission gases and sublimated iodine) are extracted during irradiation.Thus, while the Mo-99 is recovered on a batch basis, the gaseous fissionproducts can be collected on a continuous basis. As will be discussedbelow, some of the fission gases and iodine remain trapped within theNEU matrix and are recovered on a batch basis.

In this exemplary embodiment, the irradiation chamber is provided withone or more gas inlet ports 40 a and one or more gas outlet ports 40 b(shown schematically in FIG. 1). Further, provision is made for fluidcommunication between the gas inlet and outlet ports and the NEU in theirradiation chamber. An inert carrier gas (e.g., argon) is introducedinto an inlet port and circulated through the irradiation chamber whereit mixes with the gaseous fission products. The gas exiting an outletport of the irradiation chamber is subjected to a scavenging operation115 to remove the gaseous fission products before the gas isreintroduced into an inlet port of the irradiation chamber.

The gases removed by scavenging operation 115 are subjected to one ormore recovery operations 120, one of which is shown. This can be astandard chemical extraction process or a standard electrochemicalseparation. In this exemplary embodiment, it is desired to extractiodine (with I-131 often being the radioisotope of greatest interest),which can be captured in a silver zeolite trap, and the remaininggaseous fission products captured in HEPA filters for disposal. Theiodine (including I-131) is then subjected to a quality control testingoperation 125, packaged in suitable quantities (operation 127), and setfor shipment to an end user (operation 130).

References Burger_2004 (“HWVP Iodine Trap Evaluation”), Chapman 2010(“Radioactive Iodine Capture in Silver-Containing Mordenites throughNanoscale Silver Iodide Formation”), and Wang_2006 (“Simulating Gaseous¹³¹I Distribution in a Silver Zeolite Cartridge Using Sodium IodideSolution”) provide additional background for the iodine recovery.

The field of isotope extraction and separation is well developed, andMo-99 recovery process 85 could use techniques such as chemicalextraction processes and/or electrochemical separation processes. Forexample, generalized procedures for the recovery of Mo-99 from HEU havebeen developed in connection with nuclear-reactor-based operations. TheHEU is normally encapsulated in a dispersion-type target with aluminumcladding, and the HEU can take the form of mini fuel plates or pins.After irradiation (typically 10-12 days), the targets are removed fromthe reactor and cooled for several hours in the pool adjacent to thereactor before being transported to the processing hot cell.

The targets are then dissolved in nitric acid, with the possibleaddition of mercury (II) nitrate (Hg(NO.3)2) to assist the dissolutionof the aluminum. Following dissolution, the solution is fed to analumina or polymer column, and the Mo-99 is adsorbed on the column withminor amounts of other components including heavy metals. Once thecolumn is loaded with the Mo-99, the column is washed with nitric acidand then water, and then Mo-99 is stripped from the column using anammonium-hydroxide solution. Purification is carried out to remove asmuch of the heavy metals as possible. Some producers have to carry outmany purification steps in order to reduce the heavy metalconcentrations to the level to meet FDA requirements.

Chapter 2 of Reference NRC_2004 (“Medical Isotope Production withoutHighly Enriched Uranium”) provides a description ofMolybdenum-99/Technetium-99m production and use, with a description ofthe dissolution and Mo-99 recovery at pages 25-30.

In this exemplary embodiment, Mo-99 recovery process 85 uses ionicliquids, and more specifically room-temperature ionic liquids (RTILs).The recovery process includes a series of sub-processes, as will now bedescribed. Initially, the NEU (including the fission products) that isunloaded from the irradiation chamber is dissolved in an RTIL (operation135), and the Mo-99 is recovered from the solution (operation 140).Recovery operation 140, for this exemplary embodiment, entailselectrodepositing the Mo-99 onto an anode. The recovered Mo-99 is thenremoved from the anode (operation 145). For sacrificial anodes, this canentail dissolving or otherwise destroying the anode with a highercharge. In the case of a permanent anode, this can include techniquessuch as scraping.

References Pemberton_2008 (“Solubility and Electrochemistry of UranylCarbonate in a Room Temperature Ionic Liquid System”) and Pemberton_2009(“Solubility and Electrochemistry of Uranium Extracted into a RoomTemperature Ionic Liquid”) provide additional background.

The above description of the iodine recovery was somewhat simplified,and will be explained in greater detail below. In many circumstances,some of the fission gases and some of the iodine fission product remaintrapped in the NEU, and are released during Mo-99 recovery. To recoverdesired radioisotopes, provision is made to scavenge fission gases andsublimated iodine released during the Mo-99 recovery, to extract theiodine (including I-131), to subject the recovered iodine to qualitycontrol testing, to package the iodine, and to set the packaged iodinefor shipment. This is shown schematically in phantom blocks associatedwith the NEU dissolution (operation 135).

These blocks correspond generally to scavenging operation 115, recoveryoperation(s) 120, quality control testing operation 125, packagingoperation 127, and setting for shipment operation 130 that are performedduring irradiation of the NEU. While these blocks represent operationsthat are performed at different times, one or more may be implementedusing the same apparatus that is used to perform these operations duringirradiation. This is denoted by the legend “(One or more could be sharedwith irradiation chamber).” That possibility is also denotedschematically by a dashed arrow from NEU dissolution operation 135 tothe gas scavenging operation 115 that is associated with irradiating theNEU (operation 80).

Specific Embodiment with Cylindrical Irradiation Chamber with ParallelNEU Rods

FIG. 11A is a perspective view of a radioisotope generator according toa specific embodiment having a cylindrical irradiation chamber in whichare disposed parallel cylindrical NEU rods. FIG. 11B is a perspective,cutaway view of the radioisotope generator shown in FIG. 11A with thecover and the outer cylinder removed, exposing the tube assemblies,neutron reflecting regions 25 a and 25 b, and neutron generator. FIG.11C is a perspective cutaway view of the radioisotope generator shown inFIG. 11A taken from a different view to show additional details of thetube assembly and the neutron generator.

Generation and Recovery of Radioactive Iodine Isotopes Including I-131

As mentioned above, the irradiation and fission give rise to variousfission products, and some of these are in gaseous states. Radioactiveiodine 131 (sometimes referred to as ¹³¹I, radioiodine 131, or simplyI-131) is not a fission gas, but readily sublimates, and so is one ofthese gaseous fission products, and is an important radioisotope to berecovered. Embodiments of the invention are designed with the productionand recovery of I-131 and other radioactive iodine isotopes in mind.

An iodine isotope of major interest is I-131, but the fission productsinclude a number of other radioactive iodine isotopes and other elementsthat decay to radioactive iodine.

Properties and Uses of I-131

I-131 (atomic number Z=53, 78 neutrons) has a half-life of 8.02 days andis used for a variety of applications. These include diagnostic andtherapeutic thyroid applications (in either a solution or capsule form),industrial tracers, and various research applications such as antibodylabeling. I-131 is also used to label antibodies for therapeuticapplications in the treatment of cancers.

Examples of its use in radiation therapy include the treatment ofthyrotoxicosis and thyroid cancer. When a small dose of I-131 isswallowed, it is absorbed into the bloodstream in the gastrointestinal(GI) tract and concentrated from the blood by the thyroid gland, whereit begins destroying the gland's cells.

Diagnostic tests exploit the mechanism of absorption of iodine by thenormal cells of the thyroid gland. As an example I-131 is one of theradioactive isotopes of iodine that can be used to test how well thethyroid gland is functioning.

FIG. 12 shows the decay products of I-131, which is one of the primaryfission products, and constitutes on the order of 3% of the totalfission yield. In short, the I-131 decays to xenon 131, or Xe-131 (Z=54,77 neutrons), emitting a beta particle (β⁻, a gamma (γ) ray, and aneutrino (ν) in the process. The primary emissions of I-131 decay are364-keV γ rays (81% abundance) and 606-keV β-particles (89% abundance).

As shown in more detail in FIG. 12, the decay is actually a two-stepprocess where the I-131 first decays by beta decay to one of a number ofexcited states of Xe-131, emitting a β⁻ particle and a neutrino in theprocess, and Xe-131 in the excited state falls to a metastable state(Xe-131m), emitting a γ ray in the process. The first step occurs with ahalf-life of about 8 days, while the second step is, for presentpurposes, immediate.

FIG. 12 is simplified in that only two of the excited states are shown,a first that is 637 keV above the metastable state, and a second that is364 keV above the metastable state. The beta decay to the first stateresults in beta particles having a range of energies between zero and amaximum of 333 keV, while the beta decay to the second state results inbeta particles having a range of energies between zero and a maximum of606 keV. The remaining energy is carried off by the neutrino. In 82% ofthe decays to the metastable state, a 364-keV gamma ray is emitted, andin 7% of decays the decays to the metastable state, a 637-keV gamma rayis emitted. Other decay mechanisms make up the other 11% of decays tothe ground state.

The metastable isomer Xe-131m has a half-life of 11.93 days, andundergoes an isomeric transition to the stable isotope Xe-131 by themechanism of internal conversion, ejecting a single 164-keV electron inthe process. Xe-131 is one of xenon's nine stable isotopes, andconstitutes 21.2% of naturally occurring xenon.

FIG. 12 shows I-131 as a fission product, which includes some I-131nuclei which are primary fission product (fission fragments) and alsoincludes some nuclei that are decay products of other fission products.For example, fission fragments and fission products in the chaininclude:

-   -   indium 131 (In-131, Z=49, 82 neutrons), which beta decays with a        half-life of less than a second to tin 131 (Sn-131, Z=50, 81        neutrons);    -   Sn-131, which beta decays with a half-life of less than a minute        to antimony 131 (Sb-131, Z=51, 80 neutrons);    -   Sb-131, which beta decays with a half-life of 23 minutes to two        isomers of tellurium 131 (Te-131 and Te-131*, Z=52, 79        neutrons);    -   Te-131, which beta decays with a half-life of 25 minutes to        I-131 (Z=53, 78 neutrons); and    -   (Te-131*, which undergoes an isomeric transition with a        half-life on the order of 30 hours to Te-131, which beta decays        to I-131 as above).    -   The total of the I-131 fission fragments and the I-131 decay        products make up on the order of 3% of the total fission yield.

Properties and Uses of Other Radioactive Iodine Isotopes

As noted above, the fission products include a number of iodine isotopesin addition to I-131. The longer-lived radioactive fission productsinclude the following (also shown are half-lives and fission yield):

-   -   I-129 (1.59 million years, 0.54%);    -   I-131 (8.042 days, ˜3%);    -   I-132 (2.29 hours (metastable isomer 1.4 hours), 4.31%);    -   I-133 (20.8 hours (metastable isomer 9 seconds), 6.77%);    -   I-134 (52.6 minutes, 7.87%); and    -   I-135 (6.6 hours, 6.54%).

At least some of these isotopes have applications in imaging and/ormedical therapy (the most useful are believed to be I-131, I-132, andI-133).

Embodiments of the present invention also can produce radioactive iodineisotopes up to I-142. Depending on the application, it is believed thatthe radioactive iodine produced by embodiments of the present inventionwill have lower dose requirements than pure I-131 produced by othertechniques. I-130 (12.4 hours (metastable isomer 8.9 minutes)) is not afission product since it is not a fission fragment and would only beproduced in a decay chain from Te-130, except that Te-130 has ahalf-life on the order of 2.5×10²¹ years. Radioactive iodine isotopesbelow I-127 are not fission fragments and are not decay chain productssince they are blocked by stable elements.

Other radioactive iodine isotopes are short-lived (hours or minutes) andoccur in very small amounts, and can be ignored as a practical matter.I-129 accounts for 0.54% of the primary fission yields and has ahalf-life of 15.9 million years, thus being essentially stable. It ispossible to separate these radioactive iodine isotopes, and depending onthe application, there may be reasons to do so.

I-132 and I-133 are additional radioactive iodine isotopes that are ofinterest. The I-132 production scheme is as follows.

-   -   indium 132 (In-132, Z=49, 83 neutrons), which beta decays with a        half-life of less than a second to tin 132 (Sn-132, Z=50, 82        neutrons);    -   Sn-132, which beta decays with a half-life of 40 seconds to two        isomers of antimony 132 (Sb-132 and Sb-132*, Z=51, 81 neutrons);    -   Sb-132 and Sb-132*, which beta decay with respective half-lives        of 4.2 minutes and 2.8 minutes to tellurium 132 (Te-132, Z=52,        80 neutrons);    -   Te-132, which beta decays with a half-life of 3.2 days to two        isomers of iodine 132 (I-132 and I-132*, Z=53, 79 neutrons);    -   I-132, which beta decays with a half-life of 2.28 hours to xenon        132 (Xe-132, Z=54, 78 neutrons); and    -   (I-132*, which undergoes an isomeric transition with a half-life        of 1.4 hours to I-132, which beta decays to Xe-132 as above).    -   The total of the I-132 fission fragments and the I-132 decay        products make up on the order of 4.31% of the total fission        yields.

The I-133 production scheme is as follows.

-   -   indium 133 (In-133, Z=49, 84 neutrons), which beta decays with a        half-life of less than a second to tin 133 (Sn-133, Z=50, 83        neutrons);    -   Sn-133, which beta decays with a half-life of 1.4 seconds to        antimony 133 (Sb-133, Z=51, 82 neutrons);    -   Sb-133, which beta decays with a half-life of 2.5 minutes to two        isomers of tellurium 133 (Te-133 and Te-133*, Z=52, 81        neutrons);    -   Te-133, which beta decays with a half-life of 12.4 minutes to        two isomers of iodine 133 (I-133 and I-133*, Z=53, 80 neutrons);    -   (Te-133*, which undergoes an isomeric transition with a        half-life of 55.4 minutes to Te-133);    -   I-133, which beta decays with a half-life of 20.8 hours to two        isomers of xenon 133 (Xe-133 and Xe-133*, Z=54, 79 neutrons);    -   (I-133*, which undergoes an isomeric transition with a half-life        of 9 seconds to I-133, which beta decays to Xe-133 and Xe-133*        as above;    -   Xe-133, which beta decays with a half-life of 5.24 days to        cesium 133 (Cs-133, Z=55, 78 neutrons); and    -   (Xe-133*, which undergoes an isomeric transition with a        half-life of 2.19 days to Xe-133, which beta decays to Cs-133 as        above).    -   The total of the I-133 fission fragments and the I-133 decay        products make up on the order of 6.7% of the total fission        yields.

Where the end result of the iodine decay is an inert isotope of xenon(e.g., Xe-131, Xe-132, and Xe-133), there is no problem. Otherwise, theprocessing may entail additional operations. If the end result is not astable xenon isotope, it may be desirable to separate it out, forexample using electrochemical techniques or ion-exchange chromatography(ion chromatography). This would be the case for relatively long-livedradioactive substances or for undesirable stable substances such asbarium, cerium, and cesium.

Some short-lived radioactive substances can be addressed by allowing theextracted iodine additional time so the radioactive end result substancecan decay to a stable substance or a radioactive substance that issusceptible of separation. For example, I-133 decays to stable Cs-133,but I-135 and I-137 decay to radioactive cesium isotopes, which areconsidered undesirable for both imaging and therapeutic applications.

Since the irradiation cycle is on the order of 20 hours, one approach isto let the collected radioactive iodine decay for about a day (−4half-lives for I-135, and more than 1000 half-lives for I-137) so thatthe radioactive cesium can be electrochemically separated or separatedthrough ion-exchange chromatography from the iodine solution. As aresult, the resulting iodine solution would contain mainly I-127(non-radioactive), I-129, I-131, I-132, and I-133, which could be usedfor both therapeutic and imaging applications.

I-132 has a relatively short half-life—2.29 hours with an isomerictransition of I-132* of 1.39 hours. Since I-132's half-life is short,that means it decays quickly within the body, so that there is nolingering radioactivity after the procedure and the dosage is much lowerthan other iodine imaging isotopes.

Radioisotope Generator Tailored for Generation of I-131 and OtherRadioactive Iodine Isotopes

Any of the above irradiation chamber designs can be adapted to enhancethe extraction of the gaseous fission products (including I-131 andother radioactive iodine isotopes, which sublimate to a gaseous state).In particular, as mentioned above, it is desired to withdraw the gaseousfission products from the irradiation chamber during irradiation byintroducing an inert carrier gas (e.g., argon, which is inert andrelatively cheap due to its large natural occurrence), circulating itthrough the irradiation chamber to mix with the fission gases, andexhausting the gas mixture for further processing.

FIG. 13 is a simplified cross-sectional view of a radioisotope generatorsuch as that shown in the sectional view of FIG. 6 (e.g., radioisotopegenerator 10 _(sphere) or 10 _(cyl)), with additional details relatingto the circulation of gas through the irradiation chamber. Shownexplicitly are gas inlet port(s) 40 a and gas outlet port(s) 40 b. Gasventing layers 150 (shown in heavier solid lines) are provided alongsurfaces of the NEU layers, and radially extending gas venting channels155 provide gas communication paths between gas venting layers 150 andthe inlet and outlet ports.

Additional ways to increase the circulation of the carrier gas in theirradiation chamber include providing apertures in bulkheads and otherstructural elements. For embodiments using NEU in tubes, the tube wallscan be provided with holes that are generally smaller than the smallestexpected size of the NEU granules.

A pump 160 exhausts the gases from irradiation chamber 15 and the gasesare subjected to the scavenging and iodine recovery operations describedabove. The irradiation chamber is preferably maintained at a slightnegative pressure during operation.

As discussed above, some of the iodine and gaseous fission products canremain trapped in the uranium matrix, and are recovered in connectionwith the recovery of Mo-99 and other materials after the NEU is removedfrom the irradiation chamber. Providing the NEU in a granular form tendsto increase the amount of iodine and fission gases that can escape fromthe uranium matrix during irradiation and be recovered on a continuousbasis.

REFERENCES

The following references are incorporated by reference.

-   1) Burger_2004—L. L. Burger, R. D. Scheele, “HWVP Iodine Trap    Evaluation,” Pacific Northwest National Laboratory Report PNNL-14860    (September 2004)-   2) Chapman_2010—Karena W. Chapman, Peter J. Chupas, and Tina M.    Nenoff, “Radioactive Iodine Capture in Silver-Containing Mordenites    through Nanoscale Silver Iodide Formation,” J. Am. Chem. Soc., 2010,    132 (26), pp 8897-8899 (publication date (web) Jun. 15, 2010) DOI:    10.1021/ja103110y-   3) NRC_2009—“Medical Isotope Production without Highly Enriched    Uranium,” Nuclear and Radiation Studies Board, Division of Earth and    Life Studies, National Research Council of the National Academies,    The National Academies Press, Washington, D.C. (2009). Dissolution    and Mo-99 Recovery are discussed at pages 25-30.    http://www.nap.edu/openbook.php?record_id=12569-   4) Pemberton_2008—Wendy J. Pemberton, Kenneth R. Czerwinski, David    Hatchett, “Solubility and Electrochemistry of Uranyl Carbonate in a    Room Temperature Ionic Liquid System,” presented Sep. 25, 2008 in    the Radiochemistry in the Advanced Nuclear Fuel Cycle session of the    42nd Western Regional Meeting of the American Chemical Society, Las    Vegas, Nev. (Sep. 23-27, 2008)-   5) Pemberton_2009—Wendy J. Pemberton, Kenneth R. Czerwinski and    David H Hatchett, “Solubility and Electrochemistry of Uranium    Extracted into a Room Temperature Ionic Liquid,” Actinides 2009, San    Francisco, Calif., July 2009-   6) Wang 2006—Wei-Hsung Wang, Kenneth L. Matthews, I I, “Simulating    Gaseous ¹³¹I Distribution in a Silver Zeolite Cartridge Using Sodium    Iodide Solution,” Health Physics: May 2006—Volume 90-Issue 5-pp    S73-S79 DOI: 10.1097/01.HP.0000203812.30182.7b

CONCLUSION AND POTENTIAL ADVANTAGES

In conclusion it can be seen that embodiments of the present inventioncan provide safe, efficient, economical techniques for producing medicalisotopes. Embodiments of the present invention can be characterized byone or more of the following attributes, alone or in any combination:

-   -   Using neutron-reflecting material maximizes the neutron        population above the fast fission threshold of U-238 within the        NEU layer or layers, enhancing the fast fission process in the        NEU material.    -   Maintaining the neutron energy above ˜1 MeV while in the NEU        minimizes neutron capture, and hence the decay to Pu-239.    -   U-238 can be used as a primary fissionable material rather than        enriched U-235, which is used by traditional        nuclear-reactor-based methods. Depleted uranium, a byproduct        from the enrichment process that is already stored at the        Department of Energy (DOE) sites, can be utilized efficiently.        This greatly reduces the cost of Mo-99/Tc-99m production and        I-131 production due the more relaxed regulatory requirements        concerning natural uranium or depleted uranium.    -   The radioisotope generator according to embodiments of the        present invention can be widely deployed, thereby allowing        radioisotope generation closer to the end users for use as        diagnostic, therapeutic, and research medical radioisotopes in        imaging centers, hospitals, and medical research institutions.    -   Embodiments of the present invention eliminate or reduce the        need to export HEU to foreign nuclear reactors and subsequently        import radioisotopes such as Mo-99/Tc-99m and radioactive iodine        isotopes.    -   A suite of radioactive iodine radioisotopes is produced.    -   The integrated iodine dose from all the iodine radioisotopes        produced is larger than systems producing only I-131.    -   Because some of the radioisotopes have much shorter half-lives        than I-131, the produced radioisotope iodine potentially has a        broader applicability than I-131 alone—lower dose.

While the above is a complete description of specific embodiments of theinvention, the above description should not be taken as limiting thescope of the invention as defined by the claims.

1-20. (canceled)
 21. A method for producing radioisotopes comprising:introducing non-enriched uranium (“NEU”) material into a an irradiationchamber, the irradiation chamber having one or more walls formed ofneutron-reflecting material; irradiating the NEU material with neutronshaving energies above 800 keV to cause fast fission reactions to occurin the NEU material and generate fission products, wherein: at leastsome neutrons from the irradiating are reflected from at least one ofthe one or more walls, thereby increasing the path length over whichthose neutrons are in the NEU material, and the increased path lengthincreases the probability that those neutrons in the NEU material willcause fast fission reactions; and extracting the fission products fromthe NEU material.
 22. The method of claim 21 wherein one of the fissionproducts extracted comprises at least one of molybdenum-99 (Mo-99) andtechnetium-99m (Tc-99m).
 23. The method of claim 21 wherein one of thefission products extracted comprises at least one of iodine 131 (I-131)and iodine 132 (I-132).
 24. The method of claim 21 wherein the NEUmaterial in the irradiation chamber occupies a single spatiallycontiguous region.
 25. The method of claim 21 wherein the NEU materialin the irradiation chamber occupies multiple spatially disjoint regions.26. The method of claim 21 wherein the one or more walls formed ofneutron-reflecting material comprise at least one internal wall of theirradiation chamber.
 27. The method of claim 21 wherein the one or morewalls formed of neutron-reflecting material comprise an outer wall thatsurrounds all the NEU material in the irradiation chamber.
 28. Themethod of claim 21 wherein the one or more walls formed ofneutron-reflecting material comprise: at least one internal wall of theirradiation chamber; and an outer wall that surrounds all the NEUmaterial in the irradiation chamber.
 29. A method for producingradioisotopes comprising: providing a volume of NEU material;interspersing the NEU material with neutron-reflecting material;surrounding the volume of NEU material with additionalneutron-reflecting material; surrounding the additionalneutron-reflecting material with neutron-absorbing material; andirradiating the NEU material with neutrons having energies above afission threshold to cause fast fission reactions to occur in the NEUmaterial and generate fission products; wherein, for at least someneutrons, the neutron-reflecting material prolongs the time that thoseneutrons remain within the volume of NEU material, thereby increasingthe number of fast fission reactions caused by those neutrons beforethose neutrons encounter the neutron-absorbing material.
 30. The methodof claim 29, and further comprising extracting the fission products fromthe NEU material.
 31. The method of claim 30 wherein extracting at leastone of the fission products requires removing the NEU material from theirradiation chamber.
 32. The method of claim 30 wherein extracting atleast one of the fission products does not require removing the NEUmaterial from the irradiation chamber.
 33. An apparatus for producingradioisotopes comprising: a fast neutron generator; and a plurality ofspaced shells made of neutron-reflecting material, wherein: the shellsinclude an outermost shell, the shells surround the neutron generator,and the spacing between adjacent shells provides a number of regionsconfigured to receive NEU for irradiation by neutrons generated by theneutron generator.
 34. The apparatus of claim 33 wherein the outermostshell is at thicker than the remaining shell or shells.
 35. Theapparatus of claim 33, and further comprising an outer containmentvessel having one or more walls made of neutron-absorbing material toabsorb neutrons passing out of the outermost shell.
 36. The apparatus ofclaim 35 wherein the walls of the outer containment vessel are spacedfrom the outermost shell to limit the likelihood that neutrons scatteredor reflected from the walls of the outer containment vessel willencounter the outermost shell.
 37. The apparatus of claim 33 wherein theneutron generator provides neutrons having energies of at least 10 MeV.